Research and Development

Many national and international organizations conduct research and development (R&D) programmes related to ageing and long term operation. This section provides information on results of such R&D programmes.

For further information, please contact SKALTO.

IAEA Co-ordinated Research Projects (CRP)

This subsection provides information on related IAEA CRPs:

Title	Objective	Members	Conculsions/Outcomes
Ageing Management
Management of Ageing of Reactor Pressure Vessel Primary Nozzle (1993-1995)	To evaluate the actual conditions of nozzles in service; to assess structural integrity and predict the remaining life of nozzles; to report on mitigating actions (engineering and procedural) and their effectiveness in reducing the rate and effects of ageing; to compare methodologies or evaluating the current condition, for assessing structural integrity.	Bulgaria, Czech Republic, Finland, France, Germany, Hungary, India, Japan, Russia, Slovakia, Sweden, Switzerland, U.K. and USA	The final stage provided a variety of recommendations to the above mentioned tasks including effective and practical mitigation actions, for major ageing mechanisms like fatigue, brittle fracture, etc., appropriate applications of the case study results and advice on further work as well as generalization of findings from a case study on WWER-213/440 nozzle to other nozzles.
Management of Ageing of Motor Operated Isolating Valve (1993 - 1995)	To assure functionality of MOVs under both normal operating and accident conditions; to identify effective and practical methods for monitoring of MOV ageing; to develop guidelines for risk and reliability assessment of MOV ageing; to formulate MOV qualification guidelines; to establish guidelines for effective MOV maintenance.	Bulgaria, Czech Republic, Finland, France, Germany, India, Russia, Slovakia, U.K. and USA	A list of currently known monitoring methods on the MOV respectively its components produced. The report “Risk and reliability analysis” provides guidelines concerning the use of various risk and reliability engineering methods in MOV ageing analyses.The report “MOV-qualification methods and guidelines” provides a comparison of the qualification rules and methodologies existing in the countries involved in the CRP.The report “Guidelines of MOV maintenance” includes procedures for hierarchical maintenance practice.
Management of Ageing of In-containment I&C Cables Phase 1 1993 - 1995 Phase 2 1996 - 1999	Phase 1: To validate predictive cable ageing models, and to provide practical guidelines and procedures for assessing and managing the ageing of I&C cables in real plant environments. Phase 2: to resolve uncertainties in the relationship between cable monitoring techniques and DBE survivability, and provide a guide on the assessment and management of ageing in-containment cables.	Phase 1: Canada, France, Germany, India, Russian Federation, Sweden, Switzerland and U.K. Phase 2: Canada, Czech Republic, France, Germany, India, Japan, Romania, Russia, Sweden, Switzerland, U.K. and USA	Results include a summary of the relevant ageing mechanisms; operating experience for a range of NPP types; an overview of ageing management methods which are currently in use; description of cables sampling and laboratory ageing methods and of monitoring and test methods; the capabilities and the limitations of the various ageing management methods. The first output of this CRP is a pilot study TECDOC 932 “Pilot study on the management of ageing of instrumentation and control cables”. The final output is a comprehensive report, entitled TECDOC-1188 “Assessment and management of ageing of major nuclear power plant components important to safety: In-containment I&C cables Part I and Part II”.
Management of Ageing of Concrete Containment Buildings (1993 - 1995)	To compile a state-of-the-art report on concrete repair techniques and materials specifically applicable to nuclear containment structures; to develop crack mapping and acceptance/repair guidelines; to develop a set of practical condition indicators and associated guidelines for monitoring concrete containment ageing.	Canada, Czech Republic, India, Switzerland, U.K. and USA	The final output of this CRP is entitled TECDOC-1025 “Assessment and management of ageing of major nuclear power plant components important to safety: Concrete containment buildings” on the following topics: ageing management approach; ageing effects (degradations) and their causes; practical methods for detecting age-related degradation of materials/components of concrete containment structures; repair techniques and materials as well as preventive methods; a systematic ageing management programme.
Other relevant areas
Reactor Pressure Vessel Integrity	CRP-1: "Irradiation Embrittlement of Reactor Pressure Vessel Steels" CRP-2: "Analysis of the Behaviour of Advanced Reactor Pressure Vessel Steels under Neutron Irradiation" CRP-3: “Optimising Reactor Pressure Vessel Surveillance Programmes and their Analysis” CRP-4 : "Assuring Structural Integrity of Reactor Pressure Vessels" CRP-5: “Surveillance Programmes Results Application to RPV Integrity Assessment” CRP-6: “Nickel Effects in Radiation Embrittlement of RPV Materials” CRP-7: “Evaluation of Radiation damage of RPV using IAEA DB on RPV materials” CRP-8: “Master Curve Approach to monitor the Fracture Toughness of RPV in Npps” CRP-9: “Review and Benchmark of calculation methods for structural integrity assessment of RPVs during PTS		SKALTO RPV Integrity

National R & D Programmes

Many Member State organizations conduct R&D programmes related to ageing and LTO.

This subsection provides some results of these national R&D programmes. Currently only US NRC NUREG CR reports are available. Information on R&D programmes in other Member States will be added as it becomes available.

US NRC NUREG CR reports related to ageing and long term operation

NUREG CR-5643, “Insights Gained from Aging Research”

This report presents a summary of hardware-oriented engineering research programs to identify and resolve technical issues related to the aging systems, structures, and components (SSCs) in operating nuclear power plants.

NUREG CR-6428 “Effects of Thermal Aging on Fracture Toughness and Charpy–Impact Strength of Stainless Steel Pipe Welds”

The degradation of fracture toughness, tensile, and Charpy-impact properties of Type 308 stainless steel (SS) pipe welds due to thermal aging has been characterized at room temperature and 290°C. Thermal aging of SS welds results in moderate decreases in Charpy-impact strength and fracture toughness. For the various welds in this study, upper-shelf energy decreased by 50-80 J/cm2. The decrease in fracture toughness J-R curve or JIC is relatively small. Thermal aging had little or no effect on the tensile strength of the welds. Fracture properties of SS welds are controlled by the distribution and morphology of second-phase particles. Failure occurs by the formation and growth of microvoids near hard inclusions; such processes are relatively insensitive to thermal aging. The ferrite phase has little or no effect on the fracture properties of the welds. Differences in fracture resistance of the welds arise from differences in the density and size of inclusions. Mechanical-property data from the present study are consistent with results from other investigations. The existing data have been used to establish minimum expected fracture properties for SS welds.

NUREG CR-6869 “A Reliability Physics Model for Aging of Cable Insulation Materials”

This report presents a method for predicting the probability that the insulation of an aged instrumentation or control cable inside of containment will reach a critical level of embrittlement. The critical level of embrittlement can be used to support an assessment of the probability that the cable will fail to perform its function if exposed to a loss of coolant accident (LOCA). However, there are instances where cables with severely embrittled insulation have performed their function, in tests. The method predicts the probability distribution for the time it takes for the insulation of a cable subjected to a constant dose rate and temperature to reach a critical level of embrittlement. The embrittlement level is measured by the elongation at break (EAB), a condition of the cable, the greater the EAB the less the embrittlement. In order to incorporate the results in a probabilistic risk assessment, it would be necessary to estimate the probability that a cable which has reached a critical level of embrittlement would fail to perform its intended function in a LOCA.

Other International R&D Programmes

International organizations such as OECD/NEA and EC JRC also conduct R&D programmes relating to ageing and LTO.

OECD/NEA

The OECD/NEA Committee on the Safety of Nuclear Installations (CSNI) is an international committee made up of scientists and engineers. It was set up in 1973 to develop and co-ordinate the activities of the Nuclear Energy Agency concerning the technical aspects of the design, construction and operation of nuclear installations insofar as they affect the safety of such installations. The Committee’s purpose is to foster international co-operation in nuclear safety amongst the OECD Member countries.

In implementing its programme, CSNI establishes co-operative mechanisms with NEA’s Committee on Nuclear Regulatory Activities (CNRA), responsible for the activities of the Agency concerning the regulation, licensing and inspection of nuclear installations with regard to safety. It also co-operates with NEA’s Committee on Radiation Protection and Public Health and NEA’s Radioactive Waste Management Committee on matters of common interest.

The following table shows the recent CSNI programmes related to ageing and will provide links to the appropriate reports as available:

Title, Year	Specific Field	Summary
Comparison Report of RPV Pressuurised Thermal Shock International Comparative Assessment Study (PTS ICAS) NEA/CSNI R(99)3. Sep 1999	Mechanical Component: RPV PTS	The International Comparative Assessment Study of Pressurized-Thermal Shock (PTS) in Reactor Pressure Vessels was organized in 1996 to bring together an international group of experts in a comparative assessment study of integrity evaluation methods for RPVs under PTS loading. This report summarizes the recently completed International Comparative Assessment Study of Pressurized-Thermal- Chock in Reactor Pressure Vessels . (ICAS/RPV-PTS).
Prediction of Neutron Embrittlement in the Reactor Pressure Vessel. March 2000	Mechanical Component: RPV Embrittlement	The NEA Nuclear Science Committee (NSC) set up an expert group to review the state of the art modelling of radiation-induced degradation of reactor components. The NSC expert group launched two benchmarks to verify the claimed accuracies and to validate the calculation methods used, both based on the VENUS experiments performed at SCK/CEN Mol, Belgium. This reports provides detailed results from the two benchmarks.
Technical Report on Micromechanical Versus Conventional Modelling in Non-Linear Fracture Mechanics NEA/CSNI/ R(2001)6. July 2001	Mechanical Component: RPV	The CSNI Working Group on Integrity and Ageing (IAGE) created a report on micromechanical modelling to promote this technique. This report presents a comparison with non-linear fracture mechanics and highlights key aspects to better knowledge and more accurate predictions.
Technical Report on the Fatigue Crack Growth Benchmark Based on CEA Pipe Bending Tests NEA/CSNI/ R(2001)14. July 2001	Mechanical Component: Pipe	To improve estimation methods of surface crack propagation through the sickness of components.
The evaluation of Defects, Repair Criteria &Methods of Repair for Concrete Structures on Nuclear Power Plants NEA/CSNI/ R(2002)7Volume 1Volume 2. Nov 1997	Concrete Structure: NDE	To examine current practices and state of the art with regard to the evaluation of defects, repair criteria and methods of repair for concrete structures on nuclear power plants with a view to determining best practices and identifying shortfalls in current methods.
Finite Element Analysis of Ageing Reinforced and Prestressed Concrete Structures in Nuclear Plant NEA/CSNI/R(2002)13. July 2002	Concrete Structure: FEM	To summarize the need for FEA of aged concrete nuclear structures, ongoing research, assigns priorities, and identify best practices and their relative implementation costs.
Electrochemical Techniques to detect Corrosion in Concrete Structures in Nuclear Installations NEA/CSNI/R(2002)21. July 2002	Concrete Structure: NDE	To describe the electrochemical NDT that can be used in real size reinforced concrete structures to assess the corrosion condition of their reinforcement. These techniques can be used indistinctly in conventional civil engineering structures or in nuclear installations.